Provides requirements to maintain the nuclear power plant while in operation and to return the plant to service following plant outages. The rules require a mandatory program to evidence adequate safety and manage deterioration and aging effects. The rules also stipulate duties of the Authorized Nuclear Inservice Inspector to verify that the mandatory program has been completed, permitting the plant to return to service in a safe and expeditious manner. Application of this Section begins when the requirements of the construction code have been satisfied. DIVISION 2 This Division provides the requirements for the creation of the Reliability and Integrity Management (RIM) Program for advanced nuclear reactor designs. The RIM Program addresses the entire life cycle for all types of nuclear power plants, it requires a combination of monitoring, examination, tests, operation, and maintenance requirements that ensures each Structure, System, and Component (SSC) meets plant risk and reliability goals that are selected for the RIM Program.

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PDF Pages	PDF Title
5	TABLE OF CONTENTS
10	List of Sections
11	FOREWORD
13	STATEMENT OF POLICY ON THE USE OF THE ASME SINGLE CERTIFICATION MARK AND CODE AUTHORIZATION IN ADVERTISING STATEMENT OF POLICY ON THE USE OF ASME MARKING TO IDENTIFY MANUFACTURED ITEMS
14	Personnel
36	Correspondence With the Committee
38	PREFACE TO SECTION XI INTRODUCTION GENERAL
39	ORGANIZATION OF SECTION XI 1 DIVISIONS 2 ORGANIZATION OF DIVISION 1
40	3 ORGANIZATION OF DIVISION 2
41	4 REFERENCES
43	SUMMARY OF CHANGES
44	Cross-Referencing in the ASME BPVC
45	ARTICLE RIM-1 SCOPE AND RESPONSIBILITY RIM-1.1 Scope RIM-1.2 Jurisdiction RIM-1.3 Components Subject to the Requirements of This Division RIM-1.4 Owner’s Responsibility
46	RIM-1.5 Standard Units RIM-1.6 Inspection RIM-1.6.1 Duties of the Inspector and Authorized Nuclear Inservice Inspector Supervisor RIM-1.6.2 Qualification of Authorized Inspection Agencies, Inspectors, and Supervisors RIM-1.6.3 Access for Inspector RIM-1.7 Regulatory Review RIM-1.8 Tolerances
47	RIM-1.9 Referenced Standards and Specifications Tables Table RIM-1.9-1 Referenced Standards and Specifications
48	ARTICLE RIM-2 RELIABILITY AND INTEGRITY MANAGEMENT (RIM) PROGRAM RIM-2.1 RIM Program Overview RIM-2.1.1 Basis, Objective, and Process RIM-2.1.2 Responsibilities RIM-2.2 RIM Program Scope and Definition RIM-2.3 Degradation Mechanism Assessment (DMA)
49	RIM-2.4 Facility and SSC Reliability Target Allocation RIM-2.4.1 Facility-Level Risk and Reliability Targets RIM-2.4.2 SSC-Level Reliability Targets RIM-2.4.3 Scope, Level of Detail, and Technical Adequacy of the PRA RIM-2.5 Identification and Evaluation of RIM Strategies RIM-2.5.1 Identification of RIM Strategies
50	RIM-2.5.2 Evaluation of RIM Strategy Impacts on SSC Reliability RIM-2.6 Evaluation of Uncertainties RIM-2.7 RIM Program Implementation RIM-2.7.1 RIM Program Documentation RIM-2.7.2 Inservice Inspection Interval
51	RIM-2.7.3 Preservice Inspection RIM-2.7.4 Design Requirements for RIM RIM-2.7.5 Leak Detection System Requirements for RIM RIM-2.7.6 MANDE Requirements for RIM
53	RIM-2.7.7 Examination Methods and Volumes RIM-2.8 Performance Monitoring and RIM Program Updates RIM-2.9 Examination Methods RIM-2.9.1 Visual Examinations
54	RIM-2.9.2 Surface Examination RIM-2.9.3 Volumetric Examination
55	RIM-2.9.4 Alternative Examinations RIM-2.10 Additional Considerations for RIM Program Implementation RIM-2.10.1 Consequence, External Event, and Shutdown Considerations RIM-2.10.2 Principles of Risk-Informed Decision Making
57	ARTICLE RIM-3 ACCEPTANCE STANDARDS RIM-3.1 Evaluation of Examination Results and Acceptance Standards
58	ARTICLE RIM-4 REPAIR/REPLACEMENT ACTIVITIES RIM-4.1 Scope RIM-4.2 Leakage Test Requirements After a Repair/Replacement Activity RIM-4.2.1 Test Boundaries RIM-4.2.2 Gas Leakage Test RIM-4.2.3 Liquid Leakage Test
59	RIM-4.2.4 NDE in Lieu of Leakage Testing RIM-4.2.5 Exemptions From Leakage Tests RIM-4.3 Responsibilities RIM-4.4 Corrective Action RIM-4.5 Records
60	ARTICLE RIM-5 SYSTEM LEAKAGE MONITORING AND PERIODIC TESTS RIM-5.1 Scope RIM-5.2 Leakage Monitoring RIM-5.2.1 General RIM-5.2.2 Periodic Leakage Test RIM-5.3 Corrective Action RIM-5.4 Records
61	ARTICLE RIM-6 RECORDS AND REPORTS RIM-6.1 Scope RIM-6.2 Documentation Requirements RIM-6.2.1 Owner’s Responsibilities RIM-6.2.2 Owner’s Activity Report, Form OAR-1 RIM-6.2.3 Contracted Repair/Replacement Organization Responsibilities RIM-6.2.4 Owners’ Repair/Replacement Certification Record NIS-2 Responsibilities RIM-6.3 Record Retention RIM-6.3.1 Maintenance of Records RIM-6.3.2 Reproduction, Digitization, and Microfilming RIM-6.3.3 Construction Records RIM-6.3.4 RIM Program Records
62	RIM-6.3.5 Repair/Replacement Activity Records
63	ARTICLE RIM-7 GLOSSARY RIM-7.1 Terms and Definitions
65	RIM-7.2 Acronyms
66	MANDATORY APPENDIX I RIM DECISION FLOWCHARTS FOR USE WITH THE RIM PROGRAM ARTICLE I-1 FLOWCHARTS I-1.1 General
67	Figures Figure I-1.1-1 Inputs to the RIMEP for NPP Owner’s RIM Program Development
68	Figure I-1.1-2 RIM Program Development and Integration
69	Figure I-1.1-3 Process for Identifying the SSCs to Be in MANDE Program
70	Figure I-1.1-4 Selection of Strategies for SSCs to Meet Reliability Targets
71	Figure I-1.1-5 Upper Half Shows Input to MANDEEP for Developing MANDE Specification and Lower Half Shows Process for Evaluating if Section XI, Division 1 Requirements Meet MANDE Specifications
72	Figure I-1.1-6 Select, Develop, and Validate Performance Demonstration Approach to Meet SSC Reliability Target
73	MANDATORY APPENDIX II DERIVATION OF COMPONENT RELIABILITY TARGETS FROM FACILITY SAFETY REQUIREMENTS ARTICLE II-1 GENERAL REQUIREMENTS II-1.1 SCOPE II-1.2 ADEQUACY OF THE PRA II-1.3 PROCEDURE OVERVIEW
74	ARTICLE II-2 DERIVATION OF RELIABILITY TARGETS II-2.1 Facility-Level Safety Requirements II-2.2 Allocation of Reliability Targets II-2.3 Identification of Component Groups II-2.4 Trial Assignment of Reliability Targets II-2.5 Evaluation of Impacts of Reliability Targets on Facility-Level Risk II-2.6 Determination of Reliability Targets
75	MANDATORY APPENDIX III OWNER’S RECORD AND REPORT FOR RIM PROGRAM ACTIVITIES ARTICLE III-1 GUIDES TO COMPLETING FORMS III-1.1 Form OAR-1 III-1.2 Form NIS-2
76	Table III-1.1-1 Guide for Completing Form OAR-1
77	MANDATORY APPENDIX IV MONITORING AND NDE QUALIFICATION ARTICLE IV-1 INTRODUCTION IV-1.1 Scope IV-1.2 Methods IV-1.3 Owner’s Monitoring and NDE Expert Panel (MANDEEP) IV-1.3.1 General Responsibilities IV-1.3.2 MANDEEP-Specific Responsibilities
78	IV-1.3.3 MANDEEP Qualifications
79	ARTICLE IV-2 PERSONNEL QUALIFICATION IV-2.1 Basic Personnel Qualification IV-2.2 Method-Specific or Technique-Specific Personnel Qualifications IV-2.2.1 Data Acquisition Personnel IV-2.2.2 Data Evaluation Personnel
80	ARTICLE IV-3 MANDE METHODS AND TECHNIQUES RELIABILITY-BASED QUALIFICATION IV-3.1 General IV-3.2 Determination of the Qualification Requirements IV-3.3 Qualification Process IV-3.3.1 General IV-3.3.2 SSC MANDE Specifications (Figure I-1.1-5) IV-3.3.3 MANDE Technical Justification (Figure I-1.1-6) IV-3.3.4 Levels of Rigor (Figure I-1.1-6) IV-3.3.5 Qualification of NDE Methods and Techniques (Figure I-1.1-6)
81	IV-3.3.6 Monitoring Methods and Techniques (Figure I-1.1-6) IV-3.3.7 Qualification Alternatives
82	ARTICLE IV-4 MANDE PERFORMANCE DEMONSTRATIONS (FIGURE I-1.1-6) IV-4.1 General IV-4.2 Personnel Performance Demonstration for Monitoring Methods IV-4.3 NDE Personnel Performance Demonstration IV-4.4 Procedure and Equipment Performance Demonstration
83	ARTICLE IV-5 RECORDS IV-5.1 General IV-5.2 Records for Methods and Technique Qualification IV-5.3 Records for Personnel Performance Demonstrations
84	MANDATORY APPENDIX V CATALOG OF NDE REQUIREMENTS AND AREAS OF INTEREST ARTICLE V-1 EXAMINATION CATEGORIES V-1.1 Initial Consideration Table V-1.1-1 Examination Category A, Pressure-Retaining Welds in Reactor Vessels
85	Table V-1.1-2 Examination Category B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels Table V-1.1-3 Examination Category D, Full-Penetration Welded Nozzles in Vessels
86	Table V-1.1-4 Examination Category F, Pressure-Retaining Dissimilar Welds in Vessel Nozzles
87	Table V-1.1-5 Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter
88	Table V-1.1-6 Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) or Less in Diameter
89	Table V-1.1-7 Examination Category J, Pressure-Retaining Welds in Piping
90	Table V-1.1-8 Examination Category K, Welded Attachments for Vessels, Piping, Rotating Equipment, and Valves Table V-1.1-9 Examination Category L-2, Pump Casings; Examination Category M-2, Valve Bodies
91	Table V-1.1-10 Examination Category N-1, Interior of Reactor Vessels; Examination Category N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels; Examination Category N-3, Removable Core Support Structures Table V-1.1-11 Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings Table V-1.1-12 Examination Category P, All Pressure-Retaining Components
92	Table V-1.1-13 Examination Category F-A, Supports
93	MANDATORY APPENDIX VI RELIABILITY AND INTEGRITY MANAGEMENT EXPERT PANEL (RIMEP) ARTICLE VI-1 OVERVIEW VI-1.1 Responsibilities and Qualifications of RIMEP
94	MANDATORY APPENDIX VII SUPPLEMENTS FOR TYPES OF NUCLEAR REACTOR FACILITIES ARTICLE VII-1 SUPPLEMENT FOR LIGHT WATER REACTOR–TYPE FACILITIES VII-1.1 Scope VII-1.2 RIM Program — Damage Degradation Assessment VII-1.3 Acceptance Standards VII-1.3.1 Evaluation of Examination Results
95	Table VII-1.2-1 Degradation Mechanism Attributes and Attribute Criteria (LWR)
102	VII-1.3.2 Supplemental Examinations VII-1.3.3 Acceptance Standards
103	VII-1.3.4 Characterization VII-1.3.5 Acceptability VII-1.4 Acceptance Standards for Specific Examination Categories VII-1.4.1 Acceptance Standards for Examination Categories A and B, Pressure-Retaining Welds in Reactor Vessel and Other Vessels Table VII-1.3.3-1 Acceptance Standards
104	VII-1.4.2 Acceptance Standards for Examination Category D, Full-Penetration Welds of Nozzles in Vessels
105	VII-1.4.3 Acceptance Standards for Examination Category F, Pressure-Retaining Dissimilar Metal Welds in Vessel Nozzles and Category J, Pressure-Retaining Welds in Piping
106	VII-1.4.4 Acceptance Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter VII-1.4.5 Acceptance Standards for Examination Category K, Welded Attachments for Vessels, Piping, Pumps, and Valves
107	VII-1.4.6 Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter, and Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) and Less in Diameter VII-1.4.7 Acceptance Standards for Examination Categories L-2 and M-2, Equipment Casings and Valve Bodies
108	VII-1.4.8 Acceptance Standards for Examination Category N-1, Interior of Reactor Vessel; Examination Category N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels; and Examination Category N-3, Removable Core Support Structures VII-1.4.9 Acceptance Standards for Examination Category P, All Pressure-Retaining Components VII-1.4.10 Acceptable Standards for Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings
109	VII-1.4.11 Acceptance Standards for Examination Category F-A, Component Supports VII-1.5 Analytical Evaluation of Planar Flaws
110	VII-1.5.1 Acceptance Criteria for Ferritic Steel Components 4 in. (100 mm) and Greater in Thickness VII-1.5.2 Acceptance Criteria for Ferritic Components Less Than 4 in. (100 mm) in Thickness VII-1.5.3 Analytical Evaluation Procedures and Acceptance Criteria for Flaws in Austenitic and Ferritic Piping
111	VII-1.5.4 Evaluation Procedure and Acceptance Criteria for Head Penetration Nozzles of PWR Reactor Vessels
112	VII-1.6 Analytical Evaluation of Facility Operating Events VII-1.6.1 Scope VII-1.6.2 Unanticipated Operating Events VII-1.6.3 Fracture Toughness Criteria for Protection Against Failure VII-1.6.4 Operating Facility Fatigue Assessments
113	ARTICLE VII-2 SUPPLEMENT FOR LIQUID METAL REACTOR–TYPE FACILITIES
114	ARTICLE VII-3 SUPPLEMENT FOR HIGH-TEMPERATURE GAS REACTOR–TYPE FACILITIES VII-3.1 Scope VII-3.2 RIM Program — Damage Degradation Assessment VII-3.3 Acceptance Standards VII-3.3.1 Evaluation of Examination Results
115	Table VII-3.2-1 Degradation Mechanism Attributes and Attribute Criteria for High Temperature Gas Reactors
121	VII-3.3.2 Supplemental Examinations VII-3.3.3 Acceptance Standards VII-3.3.4 Characterization VII-3.3.5 Acceptability Table VII-3.3.3-1 Acceptance Standards
122	VII-3.4 Acceptance Standards for Specific Examination Categories VII-3.4.1 Acceptance Standards for Examination Categories A and B, Pressure-Retaining Welds in Reactor Vessel and Other Vessels VII-3.4.2 Acceptance Standards for Examination Category D, Full Penetration Welds of Nozzles in Vessels
123	VII-3.4.3 Acceptance Standards for Examination Category F, Pressure Dissimilar Metal Welds in Vessel Nozzles, and Examination Category J, Pressure-Retaining Welds in Piping
124	VII-3.4.4 Acceptance Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter VII-3.4.5 Acceptance Standards for Examination Category K, Welded Attachments for Vessels, Piping, Pumps, and Valves
125	VII-3.4.6 Acceptance Standards for Examination Category G-1, Pressure-Retaining Bolting Greater Than 2 in. (50 mm) in Diameter, and Examination Category G-2, Pressure-Retaining Bolting 2 in. (50 mm) and Less in Diameter VII-3.4.7 Acceptance Standards for Examination Categories L-2 and M-2, Rotating Equipment Casings and Valve Bodies
126	VII-3.4.8 Acceptance Standards for Examination Category N-1, Interior of Reactor Vessel; Examination Category N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels; and Examination Category N-3, Removable Core Support Structures VII-3.4.9 Acceptance Standards for Examination Category P, All Pressure-Retaining Components VII-3.4.10 Acceptance Standards for Examination Category O, Pressure-Retaining Welds in Control Rod Drive and Instrument Nozzle Housings
127	VII-3.4.11 Acceptance Standards for Examination Category F-A, Component Supports VII-3.5 Analytical Evaluation of Planar Flaws
128	VII-3.5.1 Acceptance Criteria for Ferritic Steel Components 4 in. (100 mm) and Greater in Thickness
129	VII-3.5.2 Acceptance Criteria for Ferritic Components Less Than 4 in. (100 mm) in Thickness VII-3.5.3 Analytical Evaluation Procedures and Acceptance Criteria for Flaws in Austenitic and Ferritic Piping
130	VII-3.5.4 Evaluation Procedure and Acceptance Criteria for Head Penetration Nozzles of Reactor Vessels VII-3.6 Analytical Evaluation of Facility Operating Events VII-3.6.1 Scope VII-3.6.2 Unanticipated Operating Events VII-3.6.3 Fracture Toughness Criteria for Protection Against Failure VII-3.6.4 Operating Facility Fatigue Assessments
132	ARTICLE VII-4 SUPPLEMENT FOR MOLTEN SALT REACTOR–TYPE FACILITIES
133	ARTICLE VII-5 SUPPLEMENT FOR GENERATION 2 LWR–TYPE FACILITIES
134	ARTICLE VII-6 SUPPLEMENT FOR FUSION MACHINE–TYPE FACILITIES
135	NONMANDATORY APPENDIX A ALTERNATE REQUIREMENTS FOR MONITORING AND NDE ARTICLE A-1 GENERAL A-1.1 Scope A-1.2 Methods A-1.3 Responsibilities
136	Figure A-1.2-1 Logic Flow Diagram of the Process for Determining Acceptability of Alternative Requirements
137	ARTICLE A-2 PROCEDURE FOR DETERMINING ACCEPTABILITY OF ALTERNATIVE MONITORING OR NDE A-2.1 Overview A-2.2 SSC Reliability Target A-2.3 Degradation Mechanisms and Failure Modes A-2.4 Approaches — Probabilistic and Deterministic
138	ARTICLE A-3 STAGE I EVALUATION A-3.1 Introduction A-3.2 Input Related to Safety Evaluation A-3.3 Input Related to Structural Evaluation A-3.4 Probabilistic Approach — Reliability Evaluation A-3.4.1 Evaluation Procedure A-3.4.2 Criteria A-3.5 Deterministic Approach — Margin Assessment A-3.5.1 Evaluation Procedure A-3.5.2 Criteria
139	ARTICLE A-4 STAGE II EVALUATION A-4.1 Introduction A-4.2 Input Related to Safety Evaluation A-4.3 Input Related to Structural Evaluation A-4.4 Detectability A-4.5 Criteria to Establish Additional Requirements
140	A-4.6 Probabilistic Approach A-4.7 Deterministic Approach
141	ARTICLE A-5 PROCEDURE FOR STRUCTURAL RELIABILITY EVALUATION FOR PASSIVE COMPONENTS A-5.1 General Requirements A-5.1.1 Scope A-5.1.2 Referenced Standards and Specifications A-5.1.3 Application A-5.2 Reliability Evaluation A-5.2.1 General A-5.2.2 Setting of Failure Scenario Figure A-5.2.1-1 Reliability Evaluation Procedure
142	A-5.2.3 Modeling A-5.2.4 Failure Probability Calculation A-5.3 Setting of Failure Scenario A-5.3.1 General A-5.3.2 Setting of Limit State A-5.3.3 Selection of Failure Modes Figure A-5.3.1-1 Procedure for Setting Failure Scenarios
143	A-5.3.4 Setting of Failure Scenarios and Evaluation Portions A-5.3.5 Setting of Reference Period A-5.4 Modeling A-5.4.1 General A-5.4.2 Formulation of Scenario Developing Process A-5.4.3 Setting of Limit-State Function A-5.4.4 Setting of Random Variables Figure A-5.4.1-1 Modeling Procedure
144	A-5.5 Reliability Calculation
145	ARTICLE A-6 RECORDS AND REPORTS A-6.1 Retention of Records And Reports
146	ARTICLE A-7 REFERENCES
147	NONMANDATORY APPENDIX B REGULATORY ADMINISTRATIVE PROVISIONS FOR NUCLEAR REACTOR FACILITIES USING RIM PROGRAM ARTICLE B-1 GENERAL REQUIREMENTS B-1.1 Scope B-1.2 Application of Code Edition B-1.3 Application of Code Cases B-1.4 Review by Regulatory and Enforcement Authority Having Jurisdiction at the Facility Site B-1.5 Report Submittal
148	ARTICLE B-2 REQUIREMENTS FOR PASSIVE COMPONENTS IN THE RIM PROGRAM
149	ENDNOTES

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ASME	2023	151


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Standard Title	ASME BPVC.XI.2-2023 Section XI, Rules for Inservice Inspection of Nuclear Reactor Facility Components, Division 2, Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Reactor Facilities
Published Code	ASME
Publication Date	2023
Pages Count	151

ASME BPVC XI 2 2023

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